Tube-To-Tubesheet Welding Development Programs to 30 Megawatt Prototype Sodium Intermediate heat Exchanger and Steam Generator (open access)

Tube-To-Tubesheet Welding Development Programs to 30 Megawatt Prototype Sodium Intermediate heat Exchanger and Steam Generator

Special welding techniques for joining Inconel weld overlays on type 316 stainless steel tubesheets and channels are described. The program for the development of the required welding procedures are divided into three specific programs: overlaying Inconel filler Metal 82 on type 316 stainless steel, welding type 316 stainless steel tubes to a type 316 stainless steel tubesheet, and welding Inconel-type 316 stainless steel composite tubes to Inconel weld overlaid type 316 tubesheets. Shock tests are described which attempt to assimilate the most drastic thermal transient that could occur in both units and to evaluate the effect of the resulting stresses on the Inconel overlay and the tube-to-tubesheet welds. (N.W.R.)
Date: 1963
Creator: Alco Products (Firm).
System: The UNT Digital Library
Test Program for 30 Megawatt Prototype Sodium Intermediate Heat Exchanger and Steam Generator (open access)

Test Program for 30 Megawatt Prototype Sodium Intermediate Heat Exchanger and Steam Generator

The designer's concept of a test program for the 30-Mw prototype intermediate heat exchanger and steam generator designed and fabricated as part of the Sodium Components Development Program is presented. The performance data will serve to verify the thermal design, or allow application of improved techniques to future designs, give an improved basis for stress analysis in design of future units, and demonstrate the capability and limitations of the units in relation to the performance specifications for which they were designed. Welding techniques for type 316 stainless steel are described. The specifications and operating conditions of the units are given along with instrumentation drawings showing test equipment design and arrangement.
Date: September 14, 1962
Creator: Alco Products (Firm).
System: The UNT Digital Library
Power Distributions for Type 3 Replacement Cores for SM-1, SM-1A and PM-2A (open access)

Power Distributions for Type 3 Replacement Cores for SM-1, SM-1A and PM-2A

Abstract: The most adverse power distribution and power peaking factors have been developed for Type 3 replacement cores for SM-1, SM-1A, and PM-2A. Also included is the 37 element Type 3 core for SM-1, which is known as SM-1 Core III and is a prototype of a Type 3 core for PM-2A. The power distributions were compared with measured distributions obtained on type 3 fuel elements in SM-1, SM-1A and PM-2A core configurations. Using the measured data as a standard, correction factors were obtained and applied to the analytical predictions of the most adverse power distribution.
Date: July 25, 1962
Creator: Oggerino, J. P.
System: The UNT Digital Library
Thermal Analysis of SM-1 Core III (open access)

Thermal Analysis of SM-1 Core III

Abstract This technical report covers the thermal analysis performed on the SM-1 Core III for both steady state and transient conditions is reported. SM-1 Core III will be used as a test for Type 3 elements in a PM-2A Core. The steady state analysis indicated minimum departure from nucleate boiling ratios (DNBR) for both design and scram conditions above the minimum criteria of 1.5. Local nucleate boiling was noted in the hot internal channels and lattice passage at scram power conditions. Loss of flow transient results indicate DNBR's above 1.5, insuring that the core is safe from burnout. Bulk boiling was noted in the hot channels and lattice passage at scram power condition.
Date: June 29, 1962
Creator: Davidson, S. L.
System: The UNT Digital Library
Hazards Report for PM-2A Core II (open access)

Hazards Report for PM-2A Core II

Abstract: This technical report describes the changes incurred in the PM-2A by the planned insertion of PM-2A Core II and the replacement of the startup and check sources. PM-2A Core II components were fabricated to specifications very nearly identical to those of PM-2A Core I. The essential difference in the cores is the boron loading which permits PM-2A Core II to meet a "one-stuck rod criteria" at beginning of life. This core has been subjected to a zero power experiment and loading procedures have been developed at the Alco Critical Facility. The nuclear and thermal and hydraulic characteristics are essentially identical to those of Core I and the replacement of the startup and check sources represent no increase in the potential for or magnitude of a hazardous situation.
Date: June 21, 1962
Creator: Coombe, John R. & Stephenson, L. D.
System: The UNT Digital Library
SM-1 Shielding Analyses (open access)

SM-1 Shielding Analyses

Abstract: This technical report analyzes gamma dose rate and neutron measurements in their relation to the SM-1 shield design and is a continuation of previous shielding measurements and analyses reported in APAE-35 and APAE-35 Supplement 2. The data reported herein are spent fuel element and rod drive pit gamma dose rates. An analysis of gamma dose rates off the core midplane is presented and compared with test data.
Date: June 20, 1962
Creator: Stephenson, L. D.
System: The UNT Digital Library
Silver - Cadmium - Indium Absorber Development (open access)

Silver - Cadmium - Indium Absorber Development

Abstract: This technical report covers development of an AG-Cd-In alternate absorber section for Army Type SM reactors. It describes the absorber material composition and the geometric configuration. It gives the nuclear and thermal analyses supporting this configuration and a detailed description of the manufacturing practice employed in fabricating the final design component.
Date: June 13, 1962
Creator: Shaw, R. A. & Harris, R. L.
System: The UNT Digital Library
Army PWR Support and Development Program Six Months Summary Report : October 1, 1961 - March 31, 1962 (open access)

Army PWR Support and Development Program Six Months Summary Report : October 1, 1961 - March 31, 1962

Abstract: Progress is reported on research and development tasks under the Program Plan for Engineering Support and Development of Army Pressurized Water Reactor Power Plants, Contract AT(30-1)-2639, during the six months' period October 1, 1061 to March 31, 1962.
Date: May 25, 1962
Creator: Dixon, M. H.
System: The UNT Digital Library
Fast Neutron Flux Measurements and Analysis in SM-1 and PM-2A Core and Vessel Mockups (open access)

Fast Neutron Flux Measurements and Analysis in SM-1 and PM-2A Core and Vessel Mockups

Abstract: This technical report contains a summary of experimental and analytical work performed on cold clean SM-1 Core II with special components and PM-2A Core I each with thermal shield and vessel mockups. The purpose of this work was to define the neutron intensity and spectra to which the pressure vessels are subjected in order to assess the effect of irradiation on the vessel lifetimes. The radial distributions of the energy dependent neutron flux per unit power in the core and vessel mockups are presented. The uncertainties associated with the measurements are also given. In addition to graphical and tabular presentations of the test results, discussions are provided on the experimental techniques and theory. A review of the analytical models is provided in addition to the calculational results. A special analytical study of the the PM-2A Core I power distribution and absolute power level determination is given. A comparison of the experimental and analytical results is made and conclusions and recommendations presented.
Date: March 31, 1962
Creator: Sontheimer, K. C.; Kemp, S. N.; Lois, L.; Clancy, E. F. & McCool, W. J.
System: The UNT Digital Library
Startup Testing of the PM-2A Nuclear Power Plant (open access)

Startup Testing of the PM-2A Nuclear Power Plant

Abstract: A detailed description of the PM-2A reactor, primary system and instrumentation is presented. A general description of the secondary system and plant arrangement at Camp Century, Greenland is included. The results of the startup physics, shielding, thermal and hydraulics and radiochemistry test are presented. the analysis performed on the shielding and thermal and hydraulic tests is included. The shielding problem at startup and the shielding modification made to the plant are described along with the radiation measurements made in these areas.
Date: March 31, 1962
Creator: McCool, W. J.; Ryan, R. M.; Richards, W. M. S.; Bradley, P. L.; Stephenson, L. D. & Cox, J. F.
System: The UNT Digital Library
PM-2A Core II Zero Power Experiment (open access)

PM-2A Core II Zero Power Experiment

Abstract: This technical report covers the zero power experiments performed on PM-2A Core II at the Alco Critical Facility. PM-2A Core II is the first replacement core for a portable pressurized reactor at Camp Century, Greenland. Core II is the same as Core I with the exception that Core II has an increased burnable poison (B-10) content. The zero power experiment consisted of fuel element uniformity test; core assembly test, development of an on-site loading procedure and an analysis of experimental data. Physical characteristics determined include distribution of fuel and B-10 in the fuel plates, minimum critical mass, control rod bank calibration, and integral rod worth. The report concludes with an analysis of the experimental data including estimated uniform and non-uniform burnup rates.
Date: March 30, 1962
Creator: Walthousen, L. D.; Mageean, J. V.; Raby, T. M. & Lee, D. H.
System: The UNT Digital Library
Summary Report of Analysis of Physics Measurements Performed on SM-1 Core I (open access)

Summary Report of Analysis of Physics Measurements Performed on SM-1 Core I

Abstract: This technical report contains a comprehensive analysis of the nuclear characteristics of the SM-1 Core I. Comparison of analytical and experimental results for neutron ages and core reactivities of a variety of cases investigated shows that the MUFT III with P-1 slowing down approximation gives the best results. At startup the core reactivity and rod bank worth under various operating conditions are investigated and compared to experiment. Core lifetime was calculated to be 16.8 MWYR compared to 16.4 MWYR experimental. The temperature coefficient has been calculated and compared to experiment as function of burnup. In Appendix A, flux distribution, temperature coefficient, effective delayed neutron fraction and core life are analyzed by Dr. R. L. Murray by one and two group modified theory series expansion calculations.
Date: March 30, 1962
Creator: Lois, L.; Paluszkiewicz, S.; Fried, B. E. & Beam, R. H.
System: The UNT Digital Library
SM-1 Reactor Vessel Closure Stud Investigation (open access)

SM-1 Reactor Vessel Closure Stud Investigation

Abstract: This report presents metallurgical .analyses and corrosion tests conducted to determine :causes of failure of two SM- 1 reactor vessel studs due to fracture in the threaded areas, after 23 months' operation (March 1959). Conclusions of corrosion tests are: (1) failure of the studs was caused by stress corrosion cracking, (2) contributing factors to failure of the studs were improper control of heat treatment of the studs and absence of treatment for stress relief after thread grinding operation.
Date: March 23, 1962
Creator: Reali, D. L.
System: The UNT Digital Library
Shielding Measurements at the SM-1 Reactor : June 1961 (open access)

Shielding Measurements at the SM-1 Reactor : June 1961

Abstract: Neutron flux and gamma radiation measurements through the SM-1 primary shield were made at the startup of Core II in June 1961. They extend previous measurements (APAE-35) both vertically and horizontally in the primary shield and in the rod drive pit. Dose rate measurements on spent fuel elements under water are also reported.
Date: March 16, 1962
Creator: Moote, F. G. & Obrist, C. H.
System: The UNT Digital Library
Hazards Report for SM-1 Core II With the SM-1 Core II With the Silver-Cadmium-Indium Control Rod Absorber Section (open access)

Hazards Report for SM-1 Core II With the SM-1 Core II With the Silver-Cadmium-Indium Control Rod Absorber Section

Abstract: In the March-April 1962 shutdown of SM-1 Core II, the SM-28 element will be re-inserted in SM-1 Core II and an SM-1 Core I element will be removed. An SM-1 Core II europium absorber will be replaced by a Ag-Cd-In absorber, and surveillance specimens will be inserted above the core support structure. Analysis of these changes concludes that re-insertion of the SM-2B stationary element and insertion of surveillance specimens do not affect hazards potential previously defined for SM-1. Replacement of the europium absorber by the Ag-Cd-In absorber will have negligible effect on reactivity control worth of the rod. The absorber meat section is encapsulated to prevent exposure of silver alloy to the primary coolant; postulated release of silver due to a cladding defect, after 2 years irradiation in SM-1, would not cause a hazard such as to restrict access to the vapor container. Possibility of steam formation in the air gap between the absorber core and cladding, causing a cladding failure, is remote. Deformation of the absorber section sufficient to cause the rod to stick, would not impair the ability of the other rods to shut down the reactor safely.
Date: March 15, 1962
Creator: Stephenson, L. D.
System: The UNT Digital Library
PL-3 Phase I Task 3 Research and Development Report (open access)

PL-3 Phase I Task 3 Research and Development Report

Abstract: This technical report summarizes the results of research and development tasks performed during Phase I of this contract and presents recommendations for future development work. Work is reported in the areas of plant assembly and relocation, housings and footings, waste heat dissipation, instrumentation, refueling systems, waste disposal, shielding, core nuclear thermal and hydraulic studies, gaseous waste processing (BWR), and critical experiments on a 5 x 5 array of Type 3 fuel elements.
Date: March 12, 1962
Creator: Humphries, G. E.
System: The UNT Digital Library
BWR Reference Design for PL-3 (open access)

BWR Reference Design for PL-3

Abstract: The natural circulation, direct cycle, boiling water reactor reference design presented in this technical report is the alternate to the preferred preliminary design developed under Phase I of the PL-3 contract. The report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and training program outline.
Date: February 28, 1962
Creator: Humphries, G. E.
System: The UNT Digital Library
Hazards Report for SM-1 Core Temperature and Flow Instrumentation (Task XIV) Covering Special Test Procedures. (open access)

Hazards Report for SM-1 Core Temperature and Flow Instrumentation (Task XIV) Covering Special Test Procedures.

Abstract: Test procedures for special tests involving in-core SM-1 temperature and flow instrumentation are described (Task XIV Package Tests). These tests involve in-core steady state flow and temperature measurements, loss of flow transients, load transients, reduced primary system pressure operations and reduced element flow. The thermal and hydraulic conditions prevailing in these tests, including steady state and transient burnout rations, are developed. The effects of reduced system pressure and flow on the burnout ratios are determined as are the expected stuck rod conditions when Task XIV test elements are installed. The effect on the maximum credible accident is included and a recommendation to conduct these Task XIV package tests is made.
Date: February 28, 1962
Creator: Bradley, P. L. & Coombe, J. R.
System: The UNT Digital Library
PWR Preliminary Design for PL-3 (open access)

PWR Preliminary Design for PL-3

Abstract: The pressurized water reactor preliminary design presented in this volume is the preferred design developed under Phase I of the PL-3 contact. This technical report presents plant design criteria, summary of plant selection, plant description, reactor and primary system description, thermal and hydraulic analysis, nuclear analysis, control and instrumentation description, shielding description, auxiliary systems, power plant equipment, waste disposal, buildings and tunnels, services, operation and maintenance, logistics, erection, cost information and a training program outline.
Date: February 28, 1962
Creator: Humphries, G. E.
System: The UNT Digital Library
Summary Report of Physics Measurements on SM-1 Core I (open access)

Summary Report of Physics Measurements on SM-1 Core I

Abstract: This technical report summarizes all core physics experiments performed on SM-1 Core I and SM-2 Rearranged and Spiked Core I throughout core life. These measurements were obtained on site during the 16.4 MWYR lifetime of SM-1 Core I and 1.6 MWYR lifetime of SM-1 Rearranged and Spiked Core I. SM-1 Core I was the first stainless steel - UO2 dispersion core to burn out in the Army Nuclear Power Program. Experimental techniques are described and a complete history of fuel movements in the core presented. Measurements include control rod bank positions during all core conditions, temperature and pressure coefficients, control rod calibrations, transient poison effects, source multiplication, and stuck rod positions. The effect of core rearrangement and spiking on core reactivity and core life is also reported. Applicable measurements from the SM-1 zero power experiments are included.
Date: February 6, 1962
Creator: Weiss, S. H.
System: The UNT Digital Library
Hazards Report for Step Transient Loading Tests on the SM-1 : Task XII (open access)

Hazards Report for Step Transient Loading Tests on the SM-1 : Task XII

Abstract: This technical report evaluates hazards involved with SM- 1 plant response and system performance tests (Task XII), to determine the response of the SM-1 plant to step electrical and steam load changes. The report describes the changes in plant equipment and operating procedures for this task and evaluates these changes for possible additional hazards to those described in APAE No. 2, Revision 1, "Hazards Summary Report for the Army Package Power Reactor SM-1."
Date: February 1, 1962
Creator: Pomeroy, D. L.
System: The UNT Digital Library
Investigation of 17-4 PH Steel Components Used in SM-1, SM-1A, PM-2A Reactor Plants (open access)

Investigation of 17-4 PH Steel Components Used in SM-1, SM-1A, PM-2A Reactor Plants

Abstract: This technical report covers the investigation of 17-4 PH steel components used in the SM-1, SM-1A and PM-2A nuclear power plants, which consist of control rod drive racks, pinions, seal shafts and associated parts. Evidence supporting the continued use of 17-4 PH steel in these applications is presented. Alco's current procurement and processing specifications for 17-4 PH steel are included. The evidence presented is in the form of stress calculations, test and environmental data and the results of non-destructive and destructive metallographic analyses.
Date: January 23, 1962
Creator: Haraway, E. S. & Saccocio, R. M.
System: The UNT Digital Library
Nuclear Measurements for Type 3 Replacement Cores for SM-1, SM-1A and PM-2A CE 3 (open access)

Nuclear Measurements for Type 3 Replacement Cores for SM-1, SM-1A and PM-2A CE 3

Abstract: This technical report contains the description and results of an experimental program to evaluate the effect of utilizing Type 3 (SM-2) replacement cores in existing Army field plants SM-1, SM-1A and PM-2A. This program, conducted at the Alco Products Critical Facility, employed SM-2 mockup fuel elements similar in composition to Type 3 fuel elements to determine start-up characteristics of Type 3 cores in SM-1, SM-1A and PM-2A core configurations. measurements include comprehensive power distribution, temperature coefficients, initial critical bank positions, control rod calibrations, critical rod configuration and material coefficients, all obtained under cold, clean, core conditions. The 45 element SM-1 and SM-1A configuration with SM-2 mockup fuel elements contain 36.4 Kg U-235 and an estimated 67.9 gm B-10, while the 37 element PM-2A configuration with SM-2 mockup elements contains 30.0 Kb U-235 and an estimated 56 gm B-10.
Date: January 11, 1962
Creator: Raby, T. M.; Walthousen, L. D.; Kemp, S. N.; McCool, W. J.; Sontheimer, K. C. & Robinson, R. A.
System: The UNT Digital Library
Criteria for Evaluating Hazards Involved in Proposed Tests On and/or Modifications To the SM-1 (open access)

Criteria for Evaluating Hazards Involved in Proposed Tests On and/or Modifications To the SM-1

Abstract: This technical report elucidates principles of hazards evaluation. The concept of hazards potential is introduced and utilized to show how a reactor system perturbation will influence its nuclear safety. Literature relating to reactor safety is referenced to provide the sources of information required for hazards analysis and show how they influence a hazards evaluation. A checklist of items which should be considered in evaluating a change, test, or modification is presented.
Date: October 18, 1961
Creator: Scoles, J. F.
System: The UNT Digital Library